Shielding benchmark calculations of selected spent fuel storage cask experiments

Conference · Fri Jan 01 04:00:00 EST 1993

OSTI ID:6935141

Broadhead, B L; Tang, J S; Parks, C V ^[1]; Taniuchi, H ^[2]

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1. Oak Ridge National Lab., TN (United States)
2. Kobe Steel Ltd. (Japan)

This paper describes the application of the three-dimensional Monte Carlo code MORSE-SGC, as implemented in the SCALE system calculational sequence SAS4, to the analysis of a series of benchmark spent fuel storage cask measurements performed at the Idaho National Engineering Laboratory. A total of five storage cask problems were analyzed to determine the expected accuracies of computational analyses using well-established Monte Carlo codes. The results presented herein represent the current status of the work. Predicted neutron dose results generally compare very favorably (within 30%) with the measurements for the cask lid, bottom, and along the cask side. Gamma-ray dose rates exhibit differing trends, depending on the measurement location. For lid and bottom doses, as well as side doses near the endfittings, agreement is again within 30%, although several exceptions are seen. However, for gamma doses along the cask side and adjacent to the active fuel, a factor of 2 overprediction is noted. Investigations into the cause of these discrepancies are currently in progress.

• Research Organization: Oak Ridge National Lab., TN (United States)
• Sponsoring Organization: DOE; EPRI; USDOE, Washington, DC (United States); Electric Power Research Inst., Palo Alto, CA (United States)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 6935141
• Report Number(s): CONF-930408-23; ON: DE93007865
• Country of Publication: United States
• Language: English

Related Subjects

052002* -- Nuclear Fuels-- Waste Disposal & Storage
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES
BENCHMARKS
CALCULATION METHODS
CASKS
COMPUTER CALCULATIONS
COMPUTER CODES
CONTAINERS
DOSE RATES
ELECTROMAGNETIC RADIATION
ENERGY SOURCES
FUELS
GAMMA RADIATION
IONIZING RADIATIONS
M CODES
MATERIALS
MONTE CARLO METHOD
NEUTRON FLUX
NUCLEAR FUELS
RADIATION FLUX
RADIATIONS
REACTOR MATERIALS
S CODES
SHIELDING
SPENT FUEL CASKS
SPENT FUELS